ML093370011
| ML093370011 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 12/14/2009 |
| From: | Beltz T Plant Licensing Branch III |
| To: | Jensen J Indiana Michigan Power Co |
| beltz T, NRR/DORL/LPL3-1, 301-415-3049 | |
| References | |
| TAC ME2165 | |
| Download: ML093370011 (4) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 14, 2009 Mr. Joseph N. Jensen Senior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106
SUBJECT:
DONALD C. COOK NUCLEAR PLANT, UNIT 1 - REQUEST FOR ADDITIONAL INFORMATION REGARDING RELIEF REQUEST (ISIR-31) (TAC NO. ME2165)
Dear Mr. Jensen:
By letter dated August 21, 2009 (Agencywide Documents Access and Management System Accession No. ML092450590), Indiana Michigan Power Company submitted a relief request for U.S. Nuclear Regulatory Commission (NRC) staff approval of an alternative to the examination requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Table IWB-2500-1, Examination Category B-D, Item No. B3.120, volumetric examinations for the CNP-1 pressurizer surge nozzle inner radius region.
The NRC staff of the Piping and NDE Branch of the Office of Nuclear Reactor Regulation completed its review of this submittal and determined that additional information was required to complete the review. The finalized requests for additional information (RAls) are being issued as an enclosure to this letter.
As agreed upon with Mr. Michael Scarpello of your staff, please respond to the RAls within 30 days of receipt of this letter. If you have any questions, please contact me at (301) 415-3049.
Sincerely, Terry A. Beltz, Senior Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-315
Enclosure:
As stated cc: Distribution via ListServ
REQUEST FOR ADDITIONAL INFORMATION RELIEF REQUEST ISIR-31 PROPOSED ALTERNATIVE FOR THE THIRD INSERVICE INSPECTION INTERVAL INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT, UNIT 1 DOCKET NO. 50-315 By letter dated August 21,2009 (Agencywide Documents Access and Management System Accession No. ML092450590), Indiana Michigan Power Company (the licensee), submitted Relief Request ISIR-31 for Nuclear Regulatory Commission review and approval. The licensee requested relief from certain examination requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). Specifically, the licensee proposed an alternative to the examination requirements of ASME Code Section XI, Table IWB-2500-1, Examination Category B-D, Item No. B3.120, volumetric examination requirements for the pressurizer inner nozzle radius (INR) section.
- 1. On pages two and three of the submittal, the licensee provides a discussion of access difficulties associated with volumetric or visual examination of the INR, 14"-1-RC-5-IRS. The configuration specifics need clarification as follows:
- a. Provide a sketch of the cast pressurizer nozzle and adjacent region showing the interferences (heaters, insulation, welds, etc,) to exterior volumetric examinations.
- b. Provide a sketch of the pressurizer nozzle region showing the restrictions (screen, batne plates, thermal sleeve, etc,) to interior visual examinations.
- c. Provide the identification of the different materials (cast carbon steel, cast stainless steel, inconel butter, stainless steel cladding, etc.) on the sketches.
- d. Provide a sketch identifying the inner radius area (percent of coverage) that can be visually examined from the interior. Discuss the process involved in moving some of these restrictions and the contribution such an effort would have on examination coverage.
- 2. Photo imaging has made great strides in camera maneuverability, miniaturization, and flaw resolution.
Provide a discussion on the state-of-the-art equipment that was evaluated for interior examinations. (Note: NUREG/CR-6943, "A Study of Remote Visual Methods to Detect Cracking in Reactor Components")
- 2
- 3. On page four of the submittal, the licensee stated that first and second inservice inspection interval examinations were performed on Units 1 and 2.
For each unit, pressurizer and interval, please identify the II\\IR examination method that was used, the examination coverage achieved, and outage date.
- 4. On page four of the submittal, the licensee provided a discussion on failure mechanisms. The INR are cladded to minimize corrosion, and the INR were examined prior to plant startup for cracks. Thermal sleeves were installed to minimize thermal cracking. NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking," documents the effects of thermal cycling on cracking the INR region.
Provide a discussion on the effects that thermal cycling has on the INR stresses and on the stress-initiating cracks.
- concurrence via memo OFFICE NRR/LPL3-1/PM NRR/LPL3-1/LA NRR/DCI/CPNB/BC NRR/LPL3-1/BC NAME TBeltz THarris TChan
- RPascarelli DATE 12/14/09 12/10/09 11/18/09 12/14/09